There are numerous types of nuclear power fuel elements. The present invention is particularly applicable to those nuclear fuel elements of the solid type which comprise a body or core of thermal neutron fissionable uranium, thorium, or mixtures thereof which may be present in an elemental state or alloyed with zirconium, niobium, or other low-cross-section, corrosion-resistant material such as stainless steel, zirconium, or zirconium alloys.
Nuclear power fuel elements generally contain two types of nuclear fuel material, both of which are valuable. It is essential that the fuel element contain a fissionable nuclear fuel material such as uranium isotopes U-233 or U-235. Fuel elements also contain nuclear fuel materials that are not originally fissionable but which can be converted to fissionable material and are, therefore, said to be fertile or potential nuclear fuel materials. For example, U-238 is a fertile material often present in fuel elements in considerable amounts. In some instances, as much as 99.3% of the uranium content may be present in the form of U-238 in the case of an unenriched element. During the course of the use of the element in a power reactor, the fissionable material, such as U-233 and U-235, releases neutrons. Some of the neutrons are trapped by the fertile but unfissionable U-238 present in the element, and the U-238 eventually becomes Pu-239 which is fissionable. In the same way, thorium, which is a fertile but unfissionable material, absorbs neutrons to become U-233 which is fissionable and useful as a nuclear fuel material. The fuel material may be in the form of a metal, an oxide, or a carbide.
Fuel elements of the solid type, with which the present invention is particularly applicable, deteriorate due to radiation damage long before the useful content of the fissionable material is used. At the same time, radioactive fission products accumulate in the fuel element. Some are gases and other are solid; however, each is objectionable in reducing the efficiency of the reactor as a whole and each exerts some part in the deterioration or reduction of the useful life of the fuel element. More particularly, many of the fission products have a high neutron capture cross section in a thermal neutron flux, thus reducing the total amounts of neutrons available for production of thermal energy or conversion of fertile material to fissile material. In addition, the gaseous fission products build up pressure within the cladding material which can result in permanent structural damage to the elements and possibly to the reactor. Since these deleterious effects occur at a time when only a fraction of the fissile values have been burned by the fission process and since the unburned fuel is too valuable to be wasted, it advantageously is reprocessed to render it fit for reuse.
None of the heretofore known methods for recovering fuel and fertile uranium or thorium from such elements has been completely satisfactory.
One method for recovering unburned fissile and fertile fuel values from solid neutron irradiated fuel elements involves dissolution of the cladding and the fuel, followed by a liquid-liquid solvent extraction process in which an aqueous nitrate feed solution containing said values is selectively extracted by contact with an organic aqueous immiscible extractant. An example of solvent extraction process for recovering uranium values, for example, is found in U.S. Pat. No. 2,848,300.
A major disadvantage of aqueous dissolution of cladding, however, is that large aqueous feed volumes containing dissolved metals must be carried through the solvent extraction process. This, in turn, leads to a large radioactive waste volume requiring expensive waste storage and handling. In addition, the solutions generally are highly corrosive and have a high fission product decay heat content. Removal of heat from aqueous waste and formation of a stable solid monolith which isolates the waste from the biosphere requires expensive processing, transportation, and disposal.
In an attempt to reduce the volume of high-level radioactive waste pollution, various other methods have been proposed, such as separately dissolving the cladding material in concentrated sulfuric acid, thus making the fuel core available for ready dissolution in a nitric acid solution. However, a cladding material, such as stainless steel, is relatively passive in sulfuric acid and, even when it does react, there is a high probability that cross contamination between the decladding solution and the core solution will result, thus further complicating the problem of recovering the fuel.
U.S. Pat. No. 2,827,405 suggests a method of desheathing fuel rods of uranium metal bars by puncturing the sheath to expose the uranium core at a plurality of points. The rod then is reacted with steam at an elevated temperature to oxidize the uranium and break the bond between the sheath and the uranium. The fuel is recovered as an oxide requiring expensive processing to convert it back to a metal.
Another method suggested in U.S. Pat. No. 2,962,371 comprises reacting the element at an elevated temperature with essentially pure anhydrous hydrogen for a time sufficient to hydride the cladding so that it falls from the core. This invention, however, is concerned with zirconium-clad fuel elements, although it is suggested that it is also applicable to elements that are clad in alloys of zirconium.
Another process for recovering the core of a zirconium-clad fuel element is disclosed in U.S. Pat. No. 3,007,769. The process comprises immersing the clad element in a substantially neutral solution of ammonium fluoride to effect the dissolution of the zirconium and separate the neutron fissionable material values from the solution.
U.S. Pat. No. 3,089,751 suggests a process for the selective separation of uranium from ferritic stainless steels. In accordance with the process disclosed therein, a nuclear fuel element consisting of a core of uranium clad in a ferritic stainless steel is heated to a temperature in the range of 850.degree. C. to 1050.degree. C. for a period of time sufficient to render the cladding susceptible to intergranular corrosion. The heated element is then cooled rapidly to a temperature range of 850.degree. C. to 615.degree. C. and then to about room temperature. The cooled element then is contacted with an aqueous nitrate solution to selectively and quantitatively dissolve the uranium from the core.
Gas phase processes for effecting the dissolution of fuel or the cladding material are disclosed in U.S. Pat. Nos. 3,149,909; 3,156,526 and 3,343,924. The problem of handling and containing gaseous fuel, however, is even greater than that for liquid phase processes.
U.S. Pat. No. 3,929,961 suggests a method of treating a nuclear fuel element enclosed in a stainless steel metal sheath which comprises disposing the fuel element with a portion thereof in an induction coil, subjecting the induction coil to a radio frequency magnetic field to induce local induction heating of the metal sheath sufficient to raise the temperature of the portion of the sheath within the coil to its melting temperature and effect local melting therein. The fuel element is moved axially relative to the induction coil within continued heating to rupture the metal sheath. The fuel values are subsequently recovered by dissolution.
It is a common disadvantage of the prior art processes that they require that all of the fuel be enriched prior to reuse or at some point require liquid or gas phase processing to enhance their fissile content with all of the problems associated therewith.